ML20009A182

From kanterella
Jump to navigation Jump to search
Forwards Util Evaluation for SEP Topic XV-20 Re Radiological Consequences of Fuel Damaging Accidents.Previous Analyses Are Acceptable & in Conformance W/Current NRC Criteria
ML20009A182
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 07/03/1981
From: Vincent R
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To: Crutchfield D
Office of Nuclear Reactor Regulation
References
TASK-15-20, TASK-RR NUDOCS 8107090158
Download: ML20009A182 (2)
v • d • e


Text

{{#Wiki_filter:,-- 4% Consumers Power Company .44 8 () 1D (L W / Bgh j ceneret offices: 212 West Michigen Avenue, Jackson, MI 49201 e (517) 788-0550 t Jul 0 j ; July 3, 1981 .) o ~; 'Sg Director, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief Operating Reactors Branch No 5 US Nuclear Regulatory Cc=rission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - SEP TOPIC XV-20, RADIOLOGICAL CONSEQUENCES OF FUEL DAMAGING ACCIDENTS Attached is the Censuners Power Company evaluation for SEP Topic XV-20 for the Big Rock Point Plant. l u f4T (C+ l Robert A Vincent Staff Licensing Engineer CC Director, Regien III, USNBC NRC Resident Inspector - Big Rock Point i l l l l l i O3r S l1 l 8107090158 810703 i PDR ADOCK 05000155 P PDR

C. .e BIG ROCK POINT PLANT SEP TOPIC XV-20, RADIOLOGICAL CONSEQUENCES C7 FUEL DAMAGING ACCIDL'TS Evaluation Analyses of fuel handling accidents for the Big Rock Point Plant have previously been performed and docketed in FHSR A=end=ent 10 and in Consuners Power Cc=pany letters to the NRC dated June 28, 1977 and Nove=ber 29, 1978. For convenience, ccpies of the FHSR a=end=ent and the referenced letters are attached. The assunptiens and =ethodolegy used for these analyses are de-scribed in Attach =ent 1 to the June 28, 1977 letter. A review of the above docu=ents was perfor=ed to verify conformance to current NRC criteria centained in SRP 15.7.h, SRP 15.7.5 and Regulatory Guide 1.25 Cenclusion The review of the a0ove docunents shewed that these analyses'are acceptable and are in ecnferrance with current NRC criteria for radiological consequences of fuel damaging accidents. l f I f I t i

WOTE: SUBSEQUEUT LEn.N CORRECTS THIS SUBMITTAL' e fr #~ %. N CODSum8m U O ~ Pomr Er/-/ Company .- s t am,s..c. iJa i s f iY e. General Of*1ces: 212 West WicNgan Avenue.Jacinson. McNgen 49201. Area Coce $17 788 CS50 June 28, 1977 Director of Nuclear Reactor Regulation Att: Mr Dc: E Davis, Acting Branch Chief Operating Reacter 3rs=ch No 2 US Nuclee. Regulatory Cc=issica Washing,,ce, DC 20555 DOCKET 50-155, LIur.sI DFR BIG ROCK PODT PLA'C - FURTEIR DATA ON PCS~ULA*ID REFUELUG ACCUEC USDI CONTAD'CC Ey letter dated April 29, 1977, Censu=ers Power Cc=pa=y was requested to provide additic:al infor=atics concerning cur sub=ittal en the rotential radielegical consequences of a postulated fuel handling acci: lent i:. side cc a --act at 3ig d Roch Feint, dated y. arch 21, 1977

  • he pur,cse of this letter is to previde the necessary respc=se. Additionally, subseque : Oc the original sub=it:al, =iner changes have bee: ::de to the location of one ra:iiation =enitor a=d Oc ala= se:

points. Therefore, our original evaluation with apprcpriate correcticas is being forwarded as Attach =en: 1. Further, the inic=ation provided in :his sub=ittal is based c= syste= design a=d installation as it vill be prior to the 1977 re-faeling operatices. Net all =odifications are presently ec=plete, but all cha.ges will be as indicated prior to refueling. l Ite: 1 l l Provide the basis for your ec clusion that the censequences of the fue' 5">~" #:s accident inside centai=ent are vell within the guidelines of 10 CyR Part 1CO when reliance is placed c:. the prepcsed au:c=atic isolation of the ve :ilction syste=. Justify your =cdel for released gasecus radioactivity =ixing within cen-eai=ent and contai=ent isclatice before the radioactivity is ec=pletely released to the envirc=ent. a. Describe the operation of the contai=ent ventilation autecatic isolation syste= and provide sche =stics of the provisiens for autc=atie isolation of the centai=ent ventilatien syste=. We understand that the aute=stic isola-tion of the centai=ent ventilation syste: vill be operative for the June 1977 and subsequent rePaelings. b. Justify the volu=e of centain=ent air in which the gasecus radicactivity released frc= a failed fuel asse=bly is assu=ed to be =ixed before release frc= the contai=ent.

  • h 0h7 focWI0

i c. Indicate the specific ventilation equip =est required to be in service during refueling that effects the =ixing of the gaseous radioactivity inside the contai=ent. d. Provi:le the exact loestic: of the two area =cnitors which win respond to the accident. Provide a sche =atic showing the surface area of the refueling cavity that each area =enitor vould be exposed to. e. Esti= ate the ti=e lapse between the contai=ent area =ocitors' response to the radicactivity frc= the da= aged fuel asse=bly and radioactivity at the purge line inteard isolatien valve. Explain is detail the reasons for the range of respcese ti=es for each area =enitor listed in your respctse of March 21, 1977 f. Provide the ti=e lapse between rece4.pt of the contai=ent isolation signal e=d cc= plate closure of the conta4--act purge line valves. g. Provide arrange =ent drawings sh=ving the relative location of the equip =ent listed in Questions 1.a. l.e and 1.d above. Restesse 1 a. The contai=ent ventilatics supply and exhaust valves are of the " fail-safe" design such t=u they vill autc=atie="y shut en less of electrical pcver or control air pressure. Un:ier operating ec=diticas, they vill s% c= a high radiatien signal fr== either of the two refaeli=g deck area =ccitors. On a high radistics alar =, a high i=pelance relay ect:act (RSS179 or 25S130) ope s, de-energi i=g coil SVX5 As a result, ec tact SVX5 cpe s a=d coils SVX1 and 57X3 are de-energi:e:i venti =g SV-9151 -9152, -9153 and -915L e=d, therefore, closing the supply a=d exhaust ve=t valves. The applicable sche =atics are en:1: sed as Attacb=en 2. This sy:e= vill be in operation for 1977 a:d subsequent refueling operations. b. As specified in the March 21, 1977 sub_ittal, two extre=es of velu=e =ixing were analy:e:1. These extre=es were: (a) No =ixing (this assu=ed a point source dose rate calculatics for =enitor respc=se) e=d (b) total rdxing (this assu=es a se=i-infinite cicud dose rate). It is rather obvicus that the actual =ixing that veuld cecur for these postulated acci:lents vould te so=ewhere between the two extre=es. Eevever, these extre=a vere chosen to indicate the verst case con:iitices cenceivable a=:1, therefore, bracket any other ec ditions. Si=ce the resultan "verst case" doses are well vithin the guidelines specified in 10 CFR 100, ve conclude that no further analysis is necessary. c. The March 21, 1977 evaluation indicated that mixing is not required to en-sure that for a postulated refueling a:cident inside ec= tai =ent eff-site doses re=ain veil below the criteria established in 10 CFS 100. Consequently, there are no require =ents for the operability of ventilation equi;=ent specifically for this purpose, and nc e are'intencied. However, there is ven.tilatien equi;=ent nor: ally in service during refueling. This equip =ent, its locatica asi descriptien, is listed in Attach =ent 3 2 l 1

8 O d. One radiatice =enitor is located on the west va],1 of the instr :=ent roc = at a position approximately T.6 =eters northvest of the reactor center line at an elevatics of 635. feet 6 inches. The second radiation =enitor is located at the southwest corner of the spent fuel storage pool, approxi-mately 6 =eters south-scutheast of the reacter center line, at an elevation of 639 feet. These loca ices are indicated in the sche =1. tic forvarded as Attach =ent k. e. Cc=plete aute=atie isolatien of the ventilatien syste= vill occur six seconis after a radistics alam set point is reached. The off-site dose vas caleu-lated by ec=bining the six-second actuatica ti=e with the applicable ti=e necessa-/ to achieve a= alar = set point condition. In order to ec=pletely bracket the "verst case" situatic=s, four separate sets of conditions were postulated. These are: (1) No Mixing - Feaking Factor = 1 5, Core Decay = 12 Ecurs (2) No Mixing - Feaking Factor = 0.6, Core Decay = 5 Days (3) Full Mixing - Fesk'ag Factor = 1.5, Core recay = 12 Hours (h) Full Mixi g - Feaking Factor = 0.6, Core Decay = 5 Days As a result of utilizing these four sets of eenditions, fcur sets of response l times were deter =ined. S.nce the actual =ixing fc,r this pestulated accide: vould 'ce expected to cecur sc=evhere between the extre=es of 0 and 100",, the actual response ti=es veuld, therefore, be bracketed between.the calculated extre=es. f. Six seconds as stipulated in our March 21 sub=ittal. g. Appropriate sche =atics are provided as Attach =ent k. I i Ite: 2 Based en the above infor=ation, the source tem pars =eters of Regulatory Guide 1.25, and the assu=ption that dropping the ftiel tra sfer eask en the cere vould da= age one-third of the core (SER dated June 1962), esti= ate the offsite deses assu=ing a postulated vorst single failure as requested in our January 17,197T letter. For the equip =ent required to reduce the ecuseque:ces of this accident, including the area===itors, provide the safety class, pcver source and t ' '" -specificatica require =ents. There should be no relia:ce on non-safety grade equip =ent to reduce exposures belev the e;uidelines of 10 CFR Part 100. Restense 2 Since the ventilation isolatien valves are fail shut in design, any lass of electrical pcver or control air pressure vill isolate centai =ent.

Further, since there are svo supply valves in series and tvo exhaust valves in series, any single failure of a valve vill still ensure full contain=ent isolatien. The contain=ent ventilatics supply and exhaust vent valves receive a " shut" signal frc= either area -ad* *'on =eniter; therefore, a failure of one =enitor vill stil'.

ensure isolation. However, assu=ing that the single failure was the failure of 2 3

r .t monitor, the verst postulated failure veuld be the failure of the closest radia-tion =cnitor. This postulated failure was fally analyzed i: our sub=ittal of March 21, 1977 As indicated previously, there have bee: =inor changes i: =c=1-ter locatics and ala::: set point.. Thus, the analysis required slight revision and is included in the Cenelusien sectica of Case I - Fuel Transfer Cask Dree of Attach =ent 1 to this sub=ittal. To su==ari:e, such a failure vould delay valve closure by a =axi=== of one second during which ti=e a thyroid dose ec==it=ent rate of 0.C6 rad /se: cud would be accu =ulated by a: individual at the site bounda y. This vould result in a= additic:al d:se ce==it=ent of 0.C6 rad to the thyroid. The resultant total whole body dose at a rate of 0.017 = rad /second would be 0.017 = rad. The current 31g Rock Point Technical Specifications allev refueling operatic:s to continue as long as one of the two area =c=1 tors is operatic a1 (ref: Technical Specificaticas Sectica 6.k.3.(3)). This conditic: could een:eive.21y lead to the re=ote possibility of having the single operatic:a1 radiatic: =cnitor fail at ths sa=e time the postulated refueling accident cecurs. Since there veuld still be ik minutes available for the operator to isolate contai =ent prior to exceeding 10 CFR 100 limits at the site bound.ary under the verti pcstulated ec=ditic=s, and si=:e a postulated a==ide:.: of this magnitude would be readily apparent to all cogni-samt persc=nel alleving a=ple ti=e for reactica, Censu=ers Fever C::pa:7 cc: eludes that no f=ther restrictic:s c: ref.:e.' '.g operations are =ecessary. The safety class and power sources for all applicable equi;=ent are 'isted in Attacb=ent 5 Ite: 3 Prepose a:y additic 11 technical specificatic s needed to ens =e that conditic s described in ite s 1 and 2 vill be =sistained (in a conservative zense) duri:s all fuel handling cperaticas within the ec= ai =ent. Restense 3 Big Rock ?cist is in the process of ec=verting its Technical Specifications into t,hc current sta:dards. This conversie: vill inecr; crate the applicable and neces-sary li=iti:g ec:ditic=s for operatics, action sta:e=ents and surveillance re-quir T *.s that pertain to these systc=s. It is fully anticipated tha: this cc versica vill be ec=plete and i=plemented by the 1978 refueling cutage. Thus, no Technical Specificaticas cha:ges are c==te= plated 2: this ti=e. David A 31xel (Si.ned) David A Bixel Nuclear Licensing Ad=inistrator CC: J0Keppler, CSU?.C . ~.

ATTAC E NT 1 Revised June 1977 EVALUA'" ION CF POTE';TIAL CO'75E0fm:0ES OF A P_Nu!UG ACCIDE;T INSIDE CONTAEDE'IT - BIG ROCK POINT PLANT CASE 1 - FUEL T*ANSTER CASE DROP Discussion Amend =en: 10, Sectie: 12 ef the Big Ecek Pei=: Final Hazards S - a / Repert (FESR) evaluated the drop of the fuel transfer cask c=to the core. "'he acci-dent was assu=ed to occur at twelve hours after shutdev: vith 225 of the fael da= aged resulting in the release of 520,000 curies of ::ble gases to contain-me=t. All haloge:s were assu=ed to be scrabbed out in the water and the cca-sequences of the accident were concluded to be insignificant in c =parison to the "=axi=u= eredible accide::." The foll: wing evaluation of this accident utilizes Regulatory Guide 1.25 ass =ptices in additien tu FESR ass==ptices considered applicable. In crder to evaluate both the area =ccitors respense ti=e and off-site c::seque=ces, calculaticas vere =ade for two cases: 1. No Mixing: The evcived a ::ivity does not dissipate appreciably and, therefore, pr:vides a point source gec=etr/ to the area =c=1:Or.* Since the activity does not =ix with the ccatai:=ent at:0 sphere, the release l rate cutside the stack is equal to the rate of activity escape frc= the reactor cavity. 2. Unifor= Mixing: "*he evolved activity dissipates and =ixes =ifer=17 vithi the ccatai==ent (free volu=e cf 2.6cE + 10 cc) to provide a se=1-infinite cicud gec=etr/ to the area== iter. No credit is taken for the ti=e requir?d f: the activity to diffase threughcut ec :airect. Since the off-site deses are directly dependent c the length of ti=e necessar*/ to isolate ce= tai =e::, both the isolatica valse closure ti=e and the area===i-tors response ti=e =ust be censidered. Assu=ptic=s -av'-#:ing the cavity re-lease rate vould :: te ecuservative for calculating the area =cciter respense time a d, therefore, =ay =0t give the =axi=u= off-site dose. != crder :: deter-1 =i=e the ec-*'~s that result i= the =axi=u= cff-site dese, both the :: =ixing and unifor= ~#v'ng cases were further subdivided. Two extre=es in radial peaking factors and two different cecurrence ti=es vere cc=bined to -=v'-*:e in 0:e case the cavity release rate and, in the other, the area =enitor response ti=e. These two subcases are: a. Accident cc:urs 12 hours after shutdev (FL*SR) and the radial peaking factor equals 15 (RG 1.25). b. Accident occurs 5 days after shu:deve and the radial peaking fae:cr equals 0.6. The folleving assu=pti::s are. applicable to all the cases evaluated:

  • Cloud dia:eter less tha: 1/3 source-te-detectoi distance.

e s 1. Fuel assumed to have an 80/20 U-235. Pa-239 fission =ixture. 2. 22% of the fuel in the core is damaged (FESR, A=endment 10). 3 Reactor operation at 240.W.g h. All gap activity frc= '..he da= aged fuel is released. This censists of 10% of all nchie gases and iodi=es i= the rods, except Kr-85 and I-129 for which 30% is assu=ed (RG 1.25). 5 A deccata=icatics factor of 100 is assu=ed for all iodices (RG 1.25). 6. Release occurs over a two-hour period (RG 1.25). 7 a. New fuel storage area monitor alar =s at 85 =r/h (located - 8 = frc: reactor versel center line). b. Spent. fuel peol area =cnitor alar =s at 160 =r/h (located - 6 = frc= reactor vessel center line). 8. The folleving =eteorologies.1 conditic:s exist per RG !.25 a. Wind speed of 1 =/s. b. Unifor= vind direction. c. A fu=igatics condition exists. The'e ass =;; ions and the.ethodolcgy described i: A;;endix_A vere used to obtain the results cemine:1 in Table I. _Cenelusiens 1. Area Menitor Response Ti=es Response times are calculated in this study to range frc= instantane:us (<< 1 second) to apprcxi=ately 15 seconds for the extre=es ec sidered. If the new fuel storage area =eniter fails fer any reasen, the spent fuel pool=== iter, located apprcxi=ately 6 =eters frc= the reacter cav-ity, vill alar = at a level of 160 =r/h. Res;ctse ti=es for this =eni:cr would range frc= instantaneous to approxi=stely 16 seconds, given the sa=e conservative bounds asst =ed for the new fuel storage area =enitor. 2. Site Soundar/ Dese The =exi=u= site boundar/ g= -a dose is conservatively calculated at 3.3 rads assu=ing stack release of all neble gases release:i to contain-ment (ie, no isolatica). Aute=atic isclation of c0ntaih=ent folleving an area== itor alar = vould : ke approxi=ately 6 ceconds. he .axi=u= site boundary thyrcid dose veuld be approxi=ately 2.L rads (7 seconds total ti=e). If the provisicc for aut =atic isolation is not available by the next refueling, approxi=stely lh =inutes veuld be available for manual isolation before 10 CFR 100 (300 Re='.li=its are reache:1. 2

CASE 2 - SI'!GI,I Et"l0!.E ORCP Discuss'en The drop of a single bundle onto the core can be evaluated using the conservative methodology of A;;endix A and the assu=;tions for the cask drop (Case 1) with the exception of the per:entage of core da= age. Ia= age of all the rods in a single bundle vould res"" da "a re ease of 1.2% of the core gap activity or 5.k% of the evcived activity in Case 1. Table II su==ari:es the effects of da= aging a single bundle. Conclusions 1. Area Mc=1 tor Response Tiras Respoue ti=es for the new fuel storage area =enitor are es1:ulated in this study to ra ge frc= approxi=ately two seccads to approxi=stely nine minutes for the extre=es ccusidered. If this =eniter fails to alar due to a mechanical or electrical failure, the spent fuel pool area =enitor response ti=es are calculated to range frc= apprcxi=stely two seec ds to apprcxi=ately five =inutes. 2. Site Soundar/ 'ose The drop of a single bundle and the resultant release of its ga; activity would =ct result in off-site deses in excess of 10 CI3100 li=its even under the assu=ption of no isclatics of e : tai:=ent. Aute=stic is01stic: folleving an area== itor ala:= results in calculated =axi=== site beundarf doses of a;;r:xi=ately 2.2 rads, thyroid and 0.252 =illirad, g*- 5 a 3

TABLE Ia Fuel Transfer Cask Cree New Fuel Storare Area Moniter No M.ixt:*e Uniform Mixin: _ Units P=1 5/T=12 h F=0.6/T=5 d P=1.5/T=12 h F=0.6/T=5 d Area.\\.cniter Resterse_ Exposure Rate ! I s (10 =) mR/h ih5 5.6 897 34.5 Response Ti=e a <1 15 u1 25 Site Bot:=ds:-r Dese a. Thyroid Dese Dese Rate ! I s Rad /s 0 35 0.06 0 35 0.c6 Total Dose Assu=ing Auto-matic Isolstica Rad 2.k 2.2 0.001 0.0007 Tots 1 Time To Reach 10 CFR 100 L1=it Min ik 79 57 >2h b. Ca==a Dose t l Dose Rate ! I s = rad /s 0.h5 0.017 0.k5 0.017 Tetal Dese Assu=ing Auto- =atic Isolation = rad 32 0.255 0.002 0.0h3 Total Dese Assu=ing !!c Isolation Rad 33 0.12 33 0.12 l

i EU I_b Fuel Transfe Cask Dree Spent Fuel Peel A et Moniter No Mixine Unife-- Mixine Units P=1.5/T=12 h P=0.6/T=5 d ?=1.5/T=12 h P=0.6/T=5 d Area Monitor Restense Exposure Rate ! I s (6 =) mR/h 233 10.0 1,k39 61.k Response Time s <1 16 <<1 2.61 Site Ronniarr Dose a. Thyroid Dose Dese Rate ! I s Raus 0 35 0.06 0.35 0.c6 Total Dose Assu=isg Auto-matic Isolatic Ead 2.k 2.5 0.001 0.0006 Total Time To Reach 10 C E 100 Li=it Min ik 79 57 >2h b. Ga==a Dose I Dose Rate ! I s =raus 0.h5 0.017 0.h5 0.017 -Total Dose l Assu=ing Auto-l matic Iscit. tic: = rad 32 0.272 0.002 0.Chh l Total Dese l Assuming :To Isolation Rad 3.3 0.12 3.,3 0.12 l l l l l ,, _,.,.,,,, ~.. ,,-,--,,v. .r.,- -, - ~.,,.,,.,.

  • o TABLE IIs g nale Bundle tre; New Fuel Storsge Ares No Mixing Uniform Mixin:

Units P=1.5/T=12 h F=0.6/T=5 d F=1.5/!=12 h F=0. e/ T= 5 d Area Moniter Restense Exposure Rate @ 1 s (10 =) =R/h 7.8 0 302 h8.5 1.86 Response Ti=e s 10 9 281 1.8 h5 7 Site Bounda:-r Oese a. Thyroid Dose Dose Rate ! I s = rad /s 18.6 3.h 18.6 3.h Total Dese Assu=ing Auto-matic Isolation = rad 31h 1,826 0.10 2.7 Total Dese Assu=ing No Isolatica (and No recay) Rad 13h 25 134 25 t. Ga==a Dose Dese Rate 6 1 s = rad /s 0.02 0.0009 0.02 0.0009 Total Dose Assu=ing Auto-0.3h 0.252 0.0001 0.Ch2 =atic Isolatic = rad Total Dose Assu=ing No Isolation (and No Decay) = rad 173 6.58 173 6.h8

_.._m TABLE iib Simle Bundle Droe Spent Fuel Peel Ares Menitor No Mixing Unifors Mixine Units P=1.5/T=12 h F=0.6/T=5 d P=1.5/T=12 h P=0.6eT=5 d Area Monitor Res cuse Exposure Rate @ 1 s (6 =) mR/h 12 5 0 5h 77.8 3 32 Response Time (160 =R/h) s 12.8 296 2.1 k8.2 Site Bounda:-r Ocse a. Thyroid Dese Dose Rate ! I s rad /s 18.6 3.h 18.6, 3.k Total Dese Assu=ing Auto- =atic Isolat':n = rad 350 2,196 0.11 3.62 l Total Dese Assu=ing No Isolation (and No Decay) Rad 13h 25 13h 25 b. Ga==a Dese Dose Rate 2 1 s nrad/s 0.02 0.0009 0.02 0.0009 Total Dose Assu=i 6 ACOC- =atic Isolation = rad-0 38 0.272 0.c001 0.0k3 Total Dose Assu=i=g No Isolatica (and No Decay) rad 173 6.kB 173 6.k8 1

APPrrDIX A Refaeli:.g Acci: lent Calculatic s A. No Mixi g Case 1. Area Mesitor Res;c se - Poi =: Scurce Gec=etry R/h = (C)(r)/s where C = reacter cavity release rate (C1/s) T = gn==a dose rate constant (R/h 31 =) s = distance to =enitor (=) 2. Site Boundary Ecse a. nyroid Dese Rads /s = (C)(3)(R)(X/Q) 1 where 3 = breathi=g raf.e (3.LTI-CL =~/s) R = adult thr:cid dose cc: version fae:cr per Reg Guide .25 (Rais/Ci i= haled) I/q = at= spherie siffusics fae::: per Fig.:re h, Reg G"

  • .25 (1.8E-Ch s/=3) b.

External 'a' cle 3cd7 Ga_=a Ocse, Se=i-hfinite Cloui Rads /s = (0.25) (iY) (C) (X/Q) where 0.2_5 = cc: version fr:= MeV to acentge: a 25 Ey = average gn=:a energy per disinteg-stics (MeV) 3. Unifc = Mixing Case 1. Contair.=en: Ac="-O sti = and Release Rate l Acen=ulatics Rate (dC/dt) = C (1 - 3) vhere 3 = contain=ent air turnover rate I I(10 ef=) (L 2

  • )I
  1. ~

0.00018/s = 2.c~or+10 cc i. l) t t i Release Rate (d /dt) = / dC/d: 0.00013 >c i I l A-1 1

2. Area Itcitor Response - Se=1-Infinite C1 cud Gec=ecr/ t R/h = (0.25) (3.600) (Iy) / dc/d o where 3.600 = conversion fres hcurs to secceds 3. Site Scundary Deze a. Thyroid Dose t Rads /s = (3) (R) (I/4) / tr/dt o b. Exter s1 ' thele 3cdy Ga==a Dose t Rads /s = (0.25) (Iy) (X/q) / dr/d o l L l I 1 1 I, 1 i I l A-2

,,a. .}F q.finik. !'t

s. _.

3 .19 ' 9:-h & 4 . m.0 r h4 p-us l 3, y ' _ r.i!;d[ g. W il i ti u lng

1 T.*

."*. - -*I g: .= - - - - Jh -.c- - - - -..- - 1-4: - -,... g. j 1: r: -;i m Il.D. .f + 3.:. tig .gl.2 M m W ll wm_ c 1:51 Li ? <,:

  • p
j. -

rl,ps; I c.! %.7,F., re i. .~ F p 1 q..o.. > .. - - - i.

.. - s.:

pa I - a -4,ga-a,,4 i .i ~.... 8 J! e l 5

.1 s. 2.* *

$ ' t.s

  1. a

.. !'s !.i M.I. li - y. ..( g- .i. l 4.. g. - ,A 't et, 1 (I L _ W,,- 5.;3,, l::,:1 D w:.s,j m2p.W jf 6-,p(l3.R l}%;!! d,g,M i.&

%g y' +.
10
- >.,

.y*4 i'- l 8 .:~~t ,. ?,f.; lg.&a. i,, Y. 3 gg. q .[o . j:

p! j
gy; ';. -

c-j -D p~-[p;k.titsh }m}i0'g !.1, ~k g ~ * - ~ ' " ~ " LQ

.y -a q (

p 9...; en y, L -~~s. , m 4 "-- ;3: lms r 1 -1, pc e !j' ,h_ QM j f f'

c. '

} s-ira 6.$.gt. . /-h~ L1 I - ! .l ii W F >.[ ' P. y; 'E a i 6 3

a. :)

l . i* J- ,.. 3 }, 4 L i ,i 3: i --4

m. !

-l 'i e 55 t :;: i-2 i .w I' ' D.r

p....}5 l

c! ll F i.. g..g.', r d ') _.2 ..y A 4 .? .I I c.: y! I '" il' F - ~ "l """$' ! 1 M '@l[.,y. I "a. ; l. 1" I 4 .W1 r 2 I

  • ( - -

D O 8! a i l 4!f .'91.- p* 4,.p. 4: ! . fit. *y 1 .l g .g th n b./ F,j I

4 'r,

.1 = != r-. b j

  • i!gp i

-+ g " g, v = y!.l !l li --s, i w j r; e i r--- i t I t -t,e g--- v.. M., -- si L1 s

  • I: 1.

4 '(*s hi, . l. -.g i g ; c 1 t

1.. !.

.'.A-sif" s

  • s 2

s I t p.. s : i 'r . s.Q-;}rs Y : -!. 1 g '8 lI.L:.s.gt1 a;st, ea 3 g..J o p a s -. .,.l yp -'h I "! sJ.c,, n.,, ' - >f I nf 4 m- ' d y :- - -.3

s..re y.*:

p r. U

s l

l w -l.S g <4 :i- 't 1 ;-

  • t..

..e po m a la i.- - u1 .i d a .o

13,.. 4

.. f _1 .i j I i ,;.d, D.* I tg k: c.! ,i, 9 6 ti D ,.ir.Q *~iiT="* "1 ::= P Md {r _._ b p .o. . - Au- . 3 i 3 .~i ...A p ,a .7 ..-6 ., I.;;g: -I 6J - , I.

61. w

!,. i 4 - 4 - t +e ,ty, l l L;r - _ _g j ll ;' m't t. j,. ._.h,]~ T '-. ..i _ c4 I II.F g) ! '. 't g..k_.'l o ' Ns

  • u.

L y '_g.f;;;=-K k.:!.;, I i ' T,ii;g i. p~; :. '!;.. f. j.' r ' [li U o.. ., M.W l: i i ip i ll, l,. p Il-{I..2- ( ',; y--I! it i. da ii h .s s i s-_ t.an..,; I, m gy ~'{" 38 F . y -e g.-. ~h 3. =. H

- II 4

3 I H 'h.t _e. -; i,, % 3 %s, = m..h '. M,. :I ? d 3 l

,. w.1, y g-:-..

r_:- ; ,, ne,,. .4 a v %s .,. ):.e 1 . t'.* i ? T, re I g.

r e

A J ?g g ;-3 "" ~ ,,g al g LA* a', 4 r. ,? .. s. -} :a+- /

3

. r. i, to ;j.

  • 1 1.

R... / e .m n 8 .t a w_, Tu r-s t i f. p 17{# M -@ r-t, i 5-. t _ ;}!N%;s q. q: /

3 'it

    • l-Qi s.

g 'Q. 113.,..:::,.- !c'- ;4--[..' ". ,i+?.*__ - D ic',,l $., -:w 6 - - - - - - -- - - - -. -..-.....T....... \\ W'lb.,.1 >_ 5',.J

  • f

.t .... 6 f,,i b.i*nn 6 l 2 <_ i .'r1 t.

I

.t c. - ! s -.,2 .,m._ _s? u. - n_5 g1.s ~1I ..... 4 ::.,.p. m l \\ .......... x...... _'t ~, - 4 *.a I e g., ,g~ i i._:. 5~' {'4,_li l j l.il! I m,. 4 ;. ! i .,.4 i,, 1. "t,, I i l g+-. . a,: .'r;.- 3 n.... i ;i i r .[ tr, 2, i} l l j. I.:..,"- .., q;. -- ,.d---'.-_.**_*l*3.* i -1 2'. E'2 -{ l .s P'- '* 2 h. 't . 'l ! 's ..,4 ~* .J.,

  • 2 r

y<.*.g. i, i,. '_7 i s_. . i 's i ~, t t .t l I [7 2! e .!;!Ifa,;.4,$' {,! ,7~s ' g ~.- - -,g - _,c....... I .?.. .4 r.a. 4 s.,. -l t' s

i. l.

2. 1. ia l I ai e :3 : 3"-- e.t. n 3 \\'S..; $.,'g:t,,# 94. g 3 2 { I e {

  • i.

l a. g ,,' s. ~* ' s,.g 4 i, a i t.i . p

aie s w e et, als surety watuts __ e _ ate tamaust watyts __s _.a se s C e es t m 0 L _ ! v.Ly g s] CV/4064 & C v /40 9 _ _. _ e m 0 6 C af f 0 4 0 Cvide ta a tv1 09 9_. _Im08C aflo se I SSLtmOa0 7 i 7,. g.: 6 S..e. s. e.. m... ' yj,

g

!.1.! .l 8 .mmm e,s..e esse: g go e ssW2 se o ..o r.70 ,.i..t. se a 3o ,o 3,, 30 si ..s a, $,$1///:.... .h l I M, " ' i s....

f t.

sv.s =:r to.ro to st o 30 gA E ~ .e g () 3' 3 3 3' 3" 3[ 3 3 Fi> R G R c i i e i i re..ese m o e.. c.. e ces ce. .s s. g! is s.: as s as is,, ess, o-c C E DDD oss? e g,, ~ g,, / PantL A ,u r.se f% o m-Tr 80 a -7,,,, ~ 'j"'8 icTs b'.. "..f.u.ap..n - ,,, C.to.s t o ... l..... ' t .~n - ctosto.n. .a.,u !,,.. so .. s u s s. waam 6 l d. e a va t,n, o-- n p a vatvs ' - as open V' is caen L' ctoses ' e.,f e., ana-mo en. it re, watwa is se

-4 3.

is etc:ao E me s,.. m l .m D my w, "a e a _,.. w s... .m ctosto war s<> go ,atv.m s .o .o i py. su. Ctosto /s.: .m... is. .sa Ihumammm. tNnt i mw, W W '8' t.'

  • ~'" av s e -

REACTCR SU'LDING VENT ISOL ATION V A LV E S o SCHEME 8501 g ,*--C u i' u aseits assies

;,*.c'.

8 "" + ~ ^v.,0. ~ i .se... .s <>e I M l m CO: I HICH R ADI ATION VENT ISOL ATION CIRCut? I SCHEME 8511 l

BIC ROCK POINT PLANT Reactor Building Ventilation Equipment Normally in farvice During Refueling EQUIPMENT DESCRIPTION CAPACITY LOCATION Supply and Exhaust 36" fans 10,000 - 14,000 cfm Outleta of 24" Vent Air Fans Cycles containment air to stack and maintains supply and exhaust containment vacuum. air lines Pipeway Coolers American Air Filter Company 3,000 - 6,000 cfm Floor elevation 616' Unit No. V2270AC Circulates and cools air within the recirc pump cavity of containment. I IIcating and American Air Filter Compar.y 6,000 cfm Floor elevations Ventilating Units Unit No. V2270AC 616' and 599' i circulates, heats and cools air within upper parts of containment (4 uni:.c) 1 e i l I M i.G L1 ta i

yG $'P S Norm \\ 4 d %-. }' I il i = ,-l - c;.,l3. /. e 7 e)>m/ m% M i 32 ( 8 ! !l gil.ia .. : 1. '.: + g5 y-

  • - 5

-la TIS i ., a I ...J' f. e It 82 jim si ;3, a i 4

r
  • 5 j;;3,,t.r.,-

O) ~!Ir , j hIMi s 3 r I 1w TiDNN'!ali b I'l I$ :E E 3 E h.,s. ! !d I h s 3 '.. 's.,. T li . m lg,. p': g) j_ s. q b.. G: R i ~ ' 2; w -s .i. 8 -4 6. it .s;f.' ... f *. . 9.I>? T" .m 4 -r%., .l.1 6 . ll l -l i ir-- 2 o

w.-

.[ +A t a i. b. / {d. }- - a - 8 L p i a. sp-s.- !, p 1 1 i.t.. I, P.. i. f.4..It.s?.T4,'; 3 5 :('.~- g./._,- t "l l - -

J 1.;

L N,1:- } . +-. a *G.. zum.p:; ;; i .4 J...

1

. er, m-.{'..... .y.. i ' - m s

  • f 11'

.i:! T' =l4 3 ydel { 8.. p s [ c...N_.,. - l 5 :,/ l s .a ,I .1 t

7. 8'
r 12 i
s-

.i. m .,q a s ;:- 10. .d.u.. -s u j il ' " " - a g m

m. ;.

3 [ '

2.....'.'... _ ),. r.---

( ' 'i T E ;T.d 0, 3, v.,,.f. [. m . b..,,A7ED I;... @.-- 0Li b h ; hy/ ;I I It'Od* 'i! W/ n :i + 1 ll .l I i ttta-II -\\r m - 11 ) i l4 3 ,6 3.LL c - E M.. ',J ':. ; _ q.g j ; i e 4h=g l s !ij -.= N f., 7'i

2. __._,ii b, Ip%p.-.1.c+1 3.c.,__.

p~ n _1_ = .a* I el i .l 1 1: - l}l.'"n I ' l-1 f :i e- [G;.3 T i .. % !.1yl2 1 7.z-- L ilit. y S! l '1 ., U n l'c L, ! -;Q)= l 1 .u s l .. m _ __A .;.7::f ij . : l: r .; i !E 3 l. %.. ! i 2 f 7 M.*t! 7 1i. .n .r i j I6I ' s!

  • i 1

3 i g

i........j

- ;;==*=, ' a

. :.,5,. _-

'w =.' 5 ~ ~ l h ~ o !a;! U. l I I u ,_,____,,i i j C*_., i f*% i ~- 7 ~. l E,. l p., ~ I

  • gId II.E l-

..... _. i.. / h *,. 9 $ M. .S 4 ng

BIC ROCK POINT PLANT Devices Currently Associated with Contair. ment Ventilation Isolation on liigh Radiation i DEVICE DESCRIPTION SAFETY CLASS LOCATION POWER S01ECE Area Monitors a) Cencral Electric Camma Detector 114B5778C4 Q-listed Reactor Terphend suspended in polystyrene scintilation Deck detector optically coupled with a photomultiplier tube. Range 1 to 1,000 mR/hr. b) Cencral Electric radiation monitor type NF01 Q-listed Control 120 V a-c*60 cps Includes d-c amplifier and trip circuit for Room Panel 1Y control roon alarm and vent valve radiation (Turbine buildirg trip modules RS 8179 and RS 8180. electrical equipment panel) RS 8179 & RS 8180 High impedance millivolt alarm trip module for vent Q-listed Control 115 V a-c 60 cps (Scheme 8511) valve high radiation contact SVX5, Hoore Industries Room Panel 1Y Hodel MVAO-10 V/X-X2/117AC. Contacts de-energized (Turbine b1dg. open. electrical equipment panel) l SVXS Vent valve high radiation trip contact. Q-listed Control 115 V a-c 60 cps (Schem 01 & 8511) Ceneral Electric relay 1211CA11J70 Room Panel 1Y l Contact de-energizes open. (Turbine b1dg. electrical equipment panel) N> aa u.

42"s, fj' sa" S Consumers AO: P0'llar KY,' C0mpally E7 m l General CMices: 2t2 West Macmagem Avenwe,Jacmeon. usem. gen 49201 e Area Coce St7 700 oS5o U November 29, 1978 Director, Nuclear Reactor Regulation Att: Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLAhT - ADDITIONAL INFORF.ATION RELATIVE TO FUEL HAhTLING ACCIDENT IN CONTAINMENT Your letter dated F.ay 22, 1978 requested Consumers Power Ccmpany to provide additional information relating to the engineered safety features which are available to mitigate the consequences of a postulated fuel handling accident inside containment (?HAIC) at the Big Rock Point Plant. The purpose of this letter is to provide the requested infor=ation. Your letter requested the isllowing four ite=s: 1. Provide system descriptions (including P& ids and control system logic and schematic diagrams) and analysis to demonstrate the extent to which existing systems required to function during the FHAIC comply with the criteria established in the Hazards Sumnary Report for engineered safety features. 2. Provide a description of the extent to which these systems will comply i with the current NRC criteria for engineered safety feature systems which are listed in the NRC Standard Review Plan, NUREG-75/087. 3. Explain why it is not necessary to have these systems meet the current NRC l criteria for ESF systems. 4. Information regarding proposed changes to the existing plant should be provided. This information should be as described in Regulatory Guide 1.70. "Y qgiwf o2 w

2 Response 1 System descriptions are provided via the following documents: 0740G30114 Sheet 2 Schematic Diagram - Air, Screen. (Attachment I) Fire and Post-Incident Systems 0740G40125 Rev P Reactor Building Ventilating, Heating (Attachment II) and Cooling System P&l Diagrams Sketch No 1 Logic Diagram for Air Supply Valves, (Attachment III) Scheme 8501 Sketch No 2 Logic Diagram for Exhaust Vent Valves, (Attachment IV) Scheme 8512 Analysis No I Written System Logic Description (Attachment V) An analysis of the existing containment isolation system shows that recent modifications to the system (ie, automatic iso 10 tion on high radiation, vacuum relief through the 24-inch supply and exhaust lites) have co'-iced the system's ability to comply with the criteria established in ene Ha:ards Summary Report. The Hazards Su= mary Report describes two 24-inch ventilatica openings; one for supply, the other for exhaust, which are closed automatically within six seconds afte r any scram signal or loss of pcwer. As described.a the above documents, these two closing features are still present with the additional feature of Ligh radiation closing also available. The Hazards Su==ary Report also describes a vacuum relief line which is intended to prevent excessive external pressure from causing da= age to the containment's integrity. The vacuum relief modification has also enhanced this system's ability to ec= ply with the vacuum relief criteria established in the FHSR by providing two independent vacuum relief lines. It is, therefore, concluded that the existing containment isolation system that is required to function during the FRAIC is in compliance with the ) criteria established in the Hazards Su= mary Report for engineered safety [ features. l i Response 2 The Containment Isolation System (CIS) is the only engineered safety feature system that is required to function to mitigate the radiological consequences of an FHAIC. An analysis of the NRC Standard Review Plan, NUREG-75/037, has shown that the following current NRC criteria need to be considered regarding the Containment Isolation System: 10 CFR Part 50 Appendix A, General Design Criteria, Criterion 23 - Protection System Failure Medes

3 10 CFR Part 50 Appendix A, General Design Criteria, Criterion 56 - Primary Containment Isolation Regulatory Guide 1.53, Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems IEEE Standard 379-1977, Standard Application of the Single Failure Criteria to Nuclear Power Generating Station Class lE Systems Criterion 56 requires that lines coanecting directly to the containment atmosphere through the containment be provided with two containment isolation valves, one inside containment, the other outside containment. The criterion further requires that valves outside containment be located as close to the containment as practical and that the automatic isolation vlaves shall be " fail-safe" upon loss of actuating power. The Big Rock Point Plant has two such lines penetrating the containment. They are the air supply and exhaust lines. Each line has two isolation valves in series (see 0740G40125, Attachment II) that are both located outside of containment, as close to the containment as practical. The redundant isolation valves are automatically controlled via electric signals controlling air to the pneumatic valve operators. The valve operators are " spring to close" which allow automatic closure on loss of air and/or electric power. The existing system meets criterion 56 with the exception of valve location. The remaining criteria concern the single failure and the failure mode. It is required that the components in the CIS have a failure =ede that is " fail-safe." It is also required that the redundant valves have control circuitry that meets the single failure criterion. As it applies to this case, the single failure criterion means tnat containment isolation shall not be prevented due to the failure of any single component in the scheme. (See schemes on Attachments III and IV.) Analysis II (Attachment VI) has been l performed to provide a single failure analysis of the containment isolation system schemes. The analysis has been performed in accordance with IEEE Standard 379-1977. Response 3 As noted in Response 2, the containment isolation system design complies with l Criterion 56 except for the location of the isolation valves. The existing design at Big Rock Point places both isolation valves outside the containment whereas the current criterica requires that one of these valves be located inside containment. The existing isolation valves are located as close to the l containment as practical. It is not credible to assume that'the conditions l inside containment during fuel handling activities could be severe enough to l compromise the physical integrity of the ventilation piping between the l containment and the first isolation valve. Consumers Power Company, I therefore, considers that the existing valve location design provides adequate assurance that Part 100 limits will not be exceeded as a result of a fuel l handling accident inside containment. l e 3... .,y_._. ___,,_y_ ,7 ,-g

4 Analysis II (Attachment VI) identifies the fact that the existing ventilation isolation system control circuitry does not meet the single failure criteria as established by IEEE Standard 379-1977. The postulated failures that do not meet single failure criteria are " stuck" contacts, mechanical failure of the solenoid valves and " hot shorts." i l The probability of each of these failures is low. The probability (per Rasmussen) of a manual switch contact failing to transpose is 1x10->/ demand. There is also a very low probability tist the contacts on a relay will stick or " weld" closed. The probability (per Rasmussen) of the solenoid valve failing to operate is lx10-3/ demand. This probability includes both electrical and mechanical failures. Since this application is concerned only with mechanical failures, and since they are less likely to occur than i electrical failures, the probability of mechanical failure is considered extremely low. The probability (per Rasmussen) of " hot shorts" or shorts to power is 1x10-8/ hour. Three events are necessary in connection with fuel handling in order to exceed Part 100 limits. First, a postulated failure of the crane is required while moving a fuel transfer cask over the reactor vessel with the head re=oved. l Second, a postulated failure of the safety brake is also required, before the l cask could possibly be dropped into the reactor. And third, the postulated failure of one of the ventilation isolation control valves, for at least fourteen minutes, is also required. The need for these three events to occur l si=ultaneously further reduces the probability of exceeding Part 100 limits. As a result of the above analysis, Consumers Power Company considers that it is not necessary to have the containment isolation system meet current NRC criteria for ESF systems in order to mitigate the radiological consequences of a fuel handling accident inside containment. Response 4 No changes to the existing plant are planned. I David A Bixel (Signed) 1 l David A Bixel I Nuclear Licensing Administrator CC: JGKeppler, USNRC l l l l [ t ,n.

8 : t l. ! f.. $ f fI 0 l p rp,.. e. !" t ll [ 3 g[3[ 3 I i

lj;t j,

g i~- !lll!!h!!dli!i4l$ph,I,1;iii!u:JdiOGili!:'li l !jt p;'l 'i:3 A id;!UID,p, I i 1 r 4;; s y !p if I !!!! 1:1.1 inid ! 1 ildit l'i;iii.li ilsu! i19 ;UI' !R i 1 F! lI l a!ji iilh iti ! c; .l l i gi; If FIIlictK.! iliill! lilll!l i m[e ,j l 4!ii H ii4JIbiti flIlhi 9 '!! iiihhii Eisl$d [.I l !diijl 0 ! fi!!k!)!lflihihIh:N$l1l!>!!b!!Ni Ib i sIqd i {n ajpe!!'. a g i;.pi!t a.h u.,i e pM'!}n!n+p'.,.i,l !i { llini m gi i

a. imp

' f: i , pap; y rp 1

saisi, l'"

I'" M: e,,1p.di,n:'d! :(,0} t,ij ih!! 8 h, hij!!![i,iii liJ{"iliii,c; k-gi"m 3 j e ! nnn o :. n;.. u r .n inau >, O [r t!!!!!Wtili H i.!!:1 :H H li! Mil:H!Iiliill!'!fil I]. } 4 4 !11 t sy r A-, -rtv ,i nm777 j 83 p i=:. d iei ? >.? i _w.i l: ': y i bI!,1 c y fk8 q

  • l~- [

l 4:y. I .as. if l i n rs Q @ @ cd y u {: j g t{_ __..7-. i Hl ly 3 fr71 H .M,, q H E .b A Lf] 3

I.j t

1;.. i_: iL I: 1 s i ( ', ' Ii i 'l l[1 T _.i {! I i.i 'd 3 31 e r u .t 4 t gi 4.- - Y'Ii4,*'.T _ l 4 I! ..%-[h. h ! I i 7744.l,$if l} '} 1.j 1.. l lI 4 If IiH,u'-:, ' :gy -,.;!.- 'J.1,-3 j tgl i, 1 i W-d/$ .j! i I I i ~ r 7 A q.i,' j... 1 .s'i -, qs,gt r, !l;j i t i -.: ; J i. j

  • a.L 3 (j sg

'i' k' i 1 j ;.i )) 'Il!,Wi ldl 'd I E t -- -- ' N ,r y) yWI Wi Wi t! ti u -,'.J M J; [;MD.#)1WiI: 4: I.,e.x1L--3..s.E;.y' 1 ' '. 4 h L.LL ,.. __. i ' sq. $- t. ). g i : I M.., 3 I M < I.s,=! e A I lll%g -r i. , ',,,:;fsy il - ; -.cm. ;,,:-,.. p ',* -

-~,,

}p i i 99 ij lf

2 i.

6 ic-- 1 i. l l, !l f..f

  • a.

.sg ^ 'I lli

! d

!I I

-w i u > n i i.a ) g g F'.),- 1 lr 5 3 '(- y-a: --- y!; $ ;:jhlf }hy }"b f .,_f*-[~i;! Ith k

p.h

,d 3, gl$.Lg*! ,l j g g g, _. _..,= = = =_-g. = 99 5 "'..*>".__?' . _.d _.pLS 1 l p i Ut j b; n. PES;;8 E,h.'ig r, j 3 ' qipi.d <yb h -w:M y.1 1gM-v % l ii,-; w 1 .m.: u ,p y 8 j 6.. 2<.:i. 7 9..i p 4 7 jik. p d ir si .l:: s 3 a 1 - I, s i .v i. i' a%.. -F M.c3g g - w-

  • a : :. p' \\ j mC

. i w.4M rod w*J '-- I1. > '.sp

b...

I' . 'd'A J /% n ~ 5 <f.:t ' 4 ..s.n.t t1:,st:..:.a.;/----d.. : I: 7 n' ' , p_, n. 4.. -~ + ai y h'>,

.---.<) f,,i,,

-i m

g.

-..,.-l'[# p1 w _.. . t[ M g, 5

;l.:3 i

l'.b T ,.8 M!, % - L,. It_#.. 1' }, i,

  • h ij. i 1 - 4

,3 ge) m., .! I s 7" g ;g.t, 'h,,,,,d,4-re1. 2 2 Q ;,i $- - iI I s c, -,g a L I I I J t ;l. t

  • m. -@ '

e s i I .I. J >y I l-- --- -. - g g

n

- k i

e. U

.i m. e i .,i 4 !~

  • L

=

I

.g d fa j $ t D !,:.'i' a. y

n.o' 4

l 5 {1 I 2. ; - l 6. ,r3,:c.,c. w e o : j .2. a,- F as 2 > h- : 4 [~ - -.--w.== 33 < [! a 1 'll as f ".Nii fe R g 2 a 1 i in ' t ? E.,l 9. :. a. 3 @ 4 '.. ' g{ l ' i l1 =l; s ' ._4 l AEl v,i r i-i.d i .a _n_. 5 m m +. r._ 8.e_1s a.. ,- J f '{ )I [' ' k8 l M-m' J k el I. ,. b~.YIJ' I 3 [m ", y,.t '!. - f ~ lI .w 6l m# la +* h- .. t} ?-} ^ (,, g-i e 6 .s I.I l E e-s .i, u% [ n, - 4.'f... _ d ;, ? *. & . l I 1 3,_ g, # [1p. =(. t 9 ---J 9 m. a

, i r

4,g__.

.., a-e --*f

'ud-- w ** ' i I,,,, t b ." 8 -} g o + .e r, ; i .r . rg 3 ..j. 'f e( <il' g"

  • l[ l r-4--.='il
  • l l-(( 7 e

y

ylN

' ff - T> ]*y1 L-- 4-6

  • Il

' 'sJ'T i "a l= g - =- ei, 4 I _l ... / *t f a t g --l 'gl',;

  • i

'i z. 1 c I h g!, 3 ', it . g..{..A 4 gi i,, h: 5 2'r='-~*t ' .W ~~ = 0 U., '- s

de 1*

m. '. ** gtt s.. 'l " I li...g!*4 i_. e. ' " ~. - ' " --"I'e..j. -

  • ,= -.. *:1 M

T.i'IQ _D; ... E. +g c g g li i g-- -;J{'tY! I:!J ?... f 4 j l .s a 'i F' e, ' (# 'l _4., t'(;)ld L el G --.h [4I.i. : 1 ' i' e ' ih, ' ' .a., .s -.y --'.!j A.4','1 1* 9,8 2' @'.6T*hj'*M.. I4 '8 y;- j-l .I t. ,".'j ; b.i :,p> -[,.,. N N. n; s2 ~ ,-}; p: 4I.3' a .c [dN -"L"h i ji Ik ._ ' II~-- ;,5 - 'I ' t 2 -, 4 M2 t* **' s' +- o e; s s-- .,.,3.i..,*.,.....

  • - m' l

,.a g -ig. 4 t ! a i-- r. 4 I rg i 1.t -" $.;..q _., .i --e q. .r g54%ti% l .-,e ; 5, s:J *... 7,y_p' M i ly,=Y [M r9ei! -. 'I >c }a m.7.t. pi ~ - m- ,o, ~ % hlj hf;[ f- ' Ib. 1 1 0,1 H -A @i-H : 2 y as' I 'l N d. _., h g 1C=E!h' h., .FI ] 2 ' -85.Al.G 9,P y J'E h .

  • m,

', ~T- . H.G ft s u!l , A .t-- 27....-~ v %.b. g.: 2 - ..: H;., oA ;; p

1. x.. + 4 - =. g.:.._.. __ril.
. i

...a : r.s,. +

  • a,=_.

r a 1 + l.e"! 3 4 i. ....q s T'-l x + i. !p .....l 3 t*b t-p t ..........v.... #." ya a...L.T_j I .-r 2

' =

a l l l'l i .^-- g.- ..).. [.. .;.... r: l b l l i.%p g ls

4.,;

.__1_}. 1 i 6 d !J -t-+- C. __..( 1 6 i l7 il .-j 3..... n..

3. g.., I ;

I I , F. t. .[ af [ .- gr o, m.-.. I .I. rii;a 3 ..i2. .L. i.. rw- >. l ', i.. 1 I

5..,

4

i =,

1 -J ,I l ) i 'e tv-e.' -it M. [, -p--- -]4 - l n. ' ~~ i;f>+ -~~ti t7i- ~~l T"" E .. 3. 9... g -- x. n., q

p..,.

i, ;a .. i...._. t p ,m,3 i g. !4!,.ini!m,.3 : #< s,.i 3 [i 9 - ~ 9ff"_T-? '. a i !!]W!.i t i j s M. o.<O M ..i, Sn f 2 .. $,l e._- -.-. --. I* I M_ h }.b #- 1l h - L_.m;b* t I m. .i w - \\ l

A?:'AC)CE:IT III Sketch No- / CalNo - Consumer $ OPER ATING SERVICES DEPARTMENT Project No ' E A r. c Power PRODUCTION & TRANSMISSION Originator *: - 1945 W. PARNALL RD Reviewed by 0cr.1pany - JACKSON MI 49203 Amned by Date 9 hah! Pagelof / N ' l

  • ej W

I ig M ll g F w e e. n O * % 2H o v o,, f J o I hb Sl* kU N h h? V / oE

  • 2 9 9*

q2-3% C C J t se & g-vc 3%o "O C 5 T 'y C 4 4 so t_- N l O 4 3 a W q w s e, a m s e ,2 a su a v u ~ 4 o c. w x 30 m

e d

QO4 t o a O *ll "2 r, 0 5 3 "I 8 5 q L %4J Q h es a oww O 8 7? E L 4 o e oc r < gtxt

v a$

L >C >U $ I$ k 3 j =; a# o l e e s 4 4 r 'p . ' l: s y -- I t 7e m 2; I A y w t, 3 4 a n 5 ._________.i__, r _ _f _ _ _ _ _ _ _. ,i .' i, 3 ,o c iv v w a dx s a;

5 i

i is e x s!. s!,i!y , Q> E- _.-- _ _ _' = 3 l a- - 4_ _ _ _.[_ 9 _ _ i t A __) ___ e__. p il s5 t a o v 3 i i a o al a m I I l ', t) + oo >, e SNd <w 4T7 g $o wa m no> eo o n g = go> zu3 5 '2 s.s d e" y r u u-w ke um P~

  • )

u ?," T N L.,% N 3 3 u. t e2, l w as 2 U O d ? ^ A n 1---~~---------'----------

3. w9 l

1 8 l P- - - T WO 2 ' b b b I j l t I W 22 0 :i I5 gi 5g I I l ds C 1 >.; T i e l =; v en e c=5 u _ _

  • _ _ _ _4.

j'_' 4 t A

  • C 2 I'

s I ~i e me 6 s -a p= i e , ei gisr ic_ i e. o ye g n y] l ,v. 46= ,N #Gt "d M l l

A"fAC}CC"' IV Sketch No I CalNo Consumers OPER ATING SERVICES DEPAP TMENT Project No eeaIe Power PRODUCTION & TRANSMISSION Originator M-m- 1945 W. PARNALL RD Reviewed by / - - Company Approved by ~ JACKSON MI 49203 Date %,h r Pa e.t.of i E I as o-g g , 'e ou ? b IN a. bc 27 r

s..E d '.

- 4 E o er e 't o u e ~ O. $E T$ N 6 < JI' g r - C v 3 %n 34 3 c 0, Q <r o e, e e e v v e r y0 0 ^ ^ W Qq % r'. ? d w < s I e a : w

c =

8 w o w

  • <Q

.c.. W c= V g - -u i _3 4,, . z ela s Q { ~! y e ~ ao 7 s s r o N Q Q N J 2 3l e 4 yW m .J o I oi e _ 3, < e-v r . i, t I o 4 I!}L U l >w > w g u t.s a-w~ GA4> f e t a se w ti l

  • i

.J ( 2 O l' a r- - - - - ' - - - l -- r r J--------- 3 >b i w i oe >J c r-s l i J I 0 I I f = J ITN C ha< Cl 3h I / = = 4 1 5 $ lb Nd 5 ( 33 u_,,__l_p__7_; u _ q _fj _ _ _ _ _ _' h w q -I g l o I lL I. T + I w j T 3r io o u,, $ S g e d< key If t! ,x ?uu a u-a z iv i hl_ _ _ _ _ _ _ _ _i E lis vg g!! e _.!__p_______________________ i______3 2 t 4 U Mo M d i Y t, 5s n 9 2=l : 31 i a ic r36_ _ _.4 _ _ _ 4 _ M o; s ce 4 g t j k u a n l 2 G5 g I =

  • t l

4 2 C0 O f N l c O to I a -a 3~ o e a .d t u6 < w-r,' A,. I J u I ee i ) g l C w zw G" om IelTo C I >c u. l - y I_ l ?; :w N li s !G l ~< S o- .i J g l N

'l o. ATTACEENT V ANALYSIS I System Logic Description I. Air Supply Valves Logic Description The Logic Diagram for the air supply valves, Scheme 8501, is shown on Sketch No 1, Attachment III. The two series air supply valves (see 0740G40125, Attachment II) to the containment are labeled CV-4096 and CV-4097. One valve is a butter'ly, the other is a check. These are pneumatically operated valves caat require air to open but are spring to close. This feature makes the isolation valves " fail-safe" on loss of instrument air. The CVs can be opened with air from either of two parallel solenoid valves labeled SV-9151 and SV-9152. Air is supplied to these solenoid valves via an instrument air line or a connection to a bank of nitrogen bottles. Power to the solenoid valves is supplied from 125 V d-c BKR #72-ID26 via parallel contacts SVXI and SVX2. The closure of either of these contacts will energize the SVs, permitting the CVs to open. SVX1 relay is energized via the proper alignment of contacts from the SS (closed under normal conditions, open during scram), HS 9001 (closed with switch in "open" or " normal after open" position), and SVX5 contact (closed on normal radiation levels, open on high). An open contact on any one of these devices provides an actuation signal that closes the CVs. The SVX2 relay is energi:ed on a v cuum relief signal which causes the SVX2 contact to close which in turn energizes the solenoid valves and opens the CVs. The relay is energized on increasing vacuum at -1.00 psig and is de-energized on decreasing vacuum at -0.70 psig. The auxiliary relays PISX1/173 and PISX2/173, as shown on 0740G30114, Attachment I, provide the necessary contacts for the desired deadband and annunciation. II. Exhaust Vent Valves Logic Description The Logic Diagram for the exhaust vent valves, Scheme 8512, is shown on Sketch No 2, Attachment IV. The logic description is identical to the one above for air supply valves with the exception of the appropriate equipment numbers and the auxiliary relays on the vacuum relief scheme. These relays are not required inasmuch as no additional annunciation is required for this scheme. oc1178-0378b-43 -.,,.-w.,.,--- an...e.,, - _.. ~,., - .--y-..,,,, -, - - - -. -v. ,,c,-

ATTACICENT VI ANAI,YSIS II I. Single Tailure Analysis for Air Supply Valves, Scheme 8501 A basic requirement in the design of a Class IE system is that no single failure of a component will interfere with the proper operation of an independent redundant counterpart or system. The redundant counterparts in this case are the series isolation valves CV-4096 and CV-4097. The point of this analysis is to determine if these valves will have the proper failure mode in the event of a single failure of a component. At this point it is appropriate to emphasize that the failure mode of the CVs is in the closed position. The closed position is required for containment isolation. The valve operator is air to open and spring to close. Independence and redundancy are the principal means of meeting the single failure criteria. The redundancy and independence of the CVs allows them to meet the single failure criteria as one set of components. Air is supplied to the CVs via parallel solenoid valves, SV-9151 and SV-9152. The solenoids are energized to supply air to the CVs to maincain them in an open position. A ccmmon type of solenoid failure vould involve having a coil wire "open" causing the solenoid to de-energize, thereby allowing the CV to " fail-safe." A less co= mon but credible type of failure would involve a mechanical failure of the core assembly that prevents the solenoid from cycling to vent the air to the control valve when the solenoid is de-energized. This type of failure would not allow the CV to close. Inasmuch as either SV can supply air i to both CVs and either SV is postulated to failure, the single failure criteria is not met at this point. The effect of interfacing systems on the CIS must also be analyzed for single failure. The instrument air supply system is one such interfacing system. Inasmuch as air is not required to perform the [ l containment isolation function, the instrument air supply does not have l to meet single failure. Relay costact SVX2 will energize the SVs on a high vacuum pressure l signal. Energizing the SVs cpens the CVs. Opening the CVs, durinr this l condition only, allows air into the containment. As the vacuum condition is eliminated the CVs will again close to mitigate the release of contaminants to the atmosphere. The relay is normally de-energized l and cannot be expected to fail electrically. A second type of frilure mode to be considered is when the relay is energized during a vacuum relief signal and then the signal is removed. It can be postalsted that 1

ATTACHMENT VI the relay contact could fail closed, thereby allowing the CVs to remain open after the vacuum is gone. If the contact'is postulated to stick closed the single failure criteria is not met. Relay SVII is normally energized to close the SVXI contact which energizes the SVs and allows the CVs to open. The relay is energi:ed through the closed contacts of the SS (Reactor Protection System), HS-9001, and relay SVX5. Opening any of these contacts causes relay SVXI to de-energize, thereby de-energi:ing the SVs and closing the CVs. The two postulated failure modes of the relay SVXI are the same as relay SVX2 above If the coil fails electrically the contact will open and circuit will " fail-safe." If the contact is postulated to stick closed, the power to the SVs would be maintained and the single failure criteria would not be met for this component. As with the above contacts, it can be postulated that the contacts for the HS-9001 or SVXS could fail in a closed position, thereby not permitting the appropriate isolation signal to de-energize SVX1 which de-energizes the SVs and allows the CVs to close. However, since these contacts represent redundant methods of providing the necessary actuation signal, a failure of either would not prevent closure of the isolation valves. The SS signal is provided through two series contacts, thereby providing redundancy and meeting the single failure criteria. The high radiation signal to the SVXS relay is provided through two series contacts from independent area monitors, thereby providing redundancy and meeting the single failure criteria. The power supply for the isolation scheme is a 125 V d-c breaker. A postulated loss of power would de-energize the SVs and allows the CVs to " fail-safe." Inasmuch as power is not required to provide containment isolation, the single failure criteria for this component is met. In summary, the single failure criteria is not met due to the configuration of the following components: SV-9151, SV-9152, SVX2 contact, and SVX1 contact. A failure of either SV to cycle properly could prevent the air to the CV from being vented and the CVs would remain open. A failure of the contacts to open in the remainder of the above components could allow the SVs to remain energized and the CVs wouli remain opeo. As a result of the lack of redundancy in the control wiring to the redundant CVs, it is also necessary to consider a postulated " hot short" in the wiring that could allow the SVs to remain or become energired independent of the ventilation isolation actuation signal. This failure also results in a nonconformance with the single failure criteria. 2

' e. ATTACEMENT VI II. Single Failure Analysis for Exhaust Vent Valves, Scheme 8512 The exact similarity between Schemes 8501 and '8512 would result in an identical analysis for both. Therefore, the analysis for Scheme 8512 will not be separately detailed. III. Conclusions In performing a systematic single failure analysis in the format suggested by IEEE Standard 379-1977, the following has been determined: 1. The required protective function is containment isolation during a postulated fuel handling accident inside containment. 2. The required protective action is the closure of the containment isolation valves. 3. The closure of the isolation valves is the only system available to provide the protective function. 4. Redundant isolation valves exist in the system, but there is no clearly defined independence or redundancy in the control circuitry that operates the isolation valves. 5. After conducting a systematic evaluation of potential failures, the single failure criteria is not met for the scheme as a whole. (Several components do, however, meet the single failure criteria separately.) The following assumptions were made in this single failure analysis: l 1. There are.o identified nondetectable failures inherit to the system design. 2. The system is qualified to withstand the affects of a seismic event without failure to any of its components. It is, therefore, not required to analyze the system in the presence of l event-caused failures and/or identified condetectable failures coincident to any single failures. 1 3

BIG ROCK PCI:IT PLA!!T - FI:!AL HA:'.ARDS St::' MARY REPORT - AME'iLE;T 10 SECTION 12 - SAFEIY ANALYSIS 1. The analysis was conducted on the basis of the following assumptions: (a) A fuel bundle of the initial core design is contained within the transfer cask. (b) All cooling water is lost frc= the transfer casr.. It is to be observed that,1f cooling water was initially present, but with a failure to replenish the supply, calculatiens indicate that it would require 90 minutes to heat all the water and transfer ~ cask frcm 100 F to 200 F, and it vould take an additional 200 minutes to boil off all this water. (c) The fuel bundle operated at apprcximately 3 4 Mut for an in-finitely long pericd prior to shutdevn. Ty.is calculated fuel bundle power recognizes either core loadings of 50 or 84 fuel bundles corresponding to reactor operation at 157 Mvt and 2LO Mut, respectively, and is maximi:cd by an appropriate radial peaking factor. (d) The elapsed time since the fuel bundle operated at power is approximately 12 hours. According to the decay heat curve used*, it may be observed that t10% of the decay heat generation rate at 12 hours covers a period rancing from 8 hcurs to 18 hours after shutdown, and it is unlikely that refueling veuld be done prior to about 12 hours afte.- shutdevn. The calculations indicate that the fuel bundle would heat up and reach fuel rod cladding failure temperatures of 1500 F about 30 minutes after cooling water is lost. Heat up of the fuel bundle vould continue until heat losses by radiation to the cask vall, conduction through the cack, and convection to the containment vessel atmosphere reached equilibrium with the decay heat generatien

  • " Fission Product Radiocetivity and Heat Generation," J. F. Stehn and E. F. Claney, A/ Conf. 15/P/1071.

~ -. = e 2 SECTION 12 - SAFEfY ANALYSIS (Contd) 1. (Contd) Calculaticas indicate that equilibrium conditions would be rate. reached after about +.r. hours following the coolant loss and that fuel temperatures at the. hot'9st point would reach abert 3000 t F. Thus no fuel pellet meltins is expected, but the fuel cladding would melt and release the 6asecus fission products contained in the fuel rod plenums. For this analysis it is assumed that the initial release of fission products frem the fuel rods occurs 30 minutes after the coolant loss and is cc=plete at the end of two hours. It is assumed that a total of 20% of the noble 6ases and halogen fission products centained in the bundle are released. Any release of solid fission products fr0m the fuel, and any additional release of fission products after the end of two houes are assumed to be insignificant in ec=parison to the indicated initial release.- Further, it is assumed that 100% of the noble gases and 50% of the halegens that are released from the fuel vould escape from the transfer cask to the centainment vessel at-mosphere. This vculd result in a containment atmosphere fission product inventory of Ic mcre than 61,000 curies of noble gases and 37,000 curies of halogens at the end of 2 hours after the coclant islostfromthetransfercas{.. This accident vould produce no attendant pressure effects inside the containment vessel and there-fore no uncontrolled leckage to the environs. The existence of high radiation levels vould be indicated and alarmed by the con-l i tain=ent vessel area monitors and by the stack moniter. Consequently, an operator would be able to control the subsequent release to the l environs in accordance with the permissible annual average stack emission rates. These consequences are insignificant in ec=parison O to the consequences described for the " maximum credible accident." l 2. The analysis assumes that the crane cable er both cask slings break j vhen the transfer cask is directly over the core at an elevation one i foot above the floor level. It is further acce.med that the refueling l l l 1

.s SECTION 12 - SAFETf ANALYSIS (Contd) 2. (Contd) platfor= and water do not retard the free fall of the cask and that the cask hits the core structure with its smallest frontal area and penetrates it in a vertical path. Calculated speed. for the cask when it contacts the top guide is 38 ft/sec. The eer-responding kinetic energy is 1.1 x 10 ft-lb. This energy would be absorbed in bending, shearing and ec= pressing the top guide, support tubes and channels and fuel. The contrcl rod veuld crush along with the fuel channels since the index tube and drive mechanis: have =uch greater strength. It is calculated that the total energy vould be dissipated when the support tubes are cc=- pressed and two feet of the channels and fuel are ec= pressed. The cask.vould come to rest abcut five feet above the botte of the reactor vessel leaving the vessel intact. In this condition there vould no longer be optimum core neutron =cderation. The neutron absorption characteristics of the cask and increased leakage vould also contribute to the creation of a lever core multipliestion constant. The cross-sectional area of the cask is such th'at atcut 22% cf the fuel in the 2h0 Mvt rated core (which slightly exceeds the pr0-portionate1/3cfthe157Mvtratedcore)couldbedamaged. This would result in the release of gasecus fission products frc= the damaged fuel rods. The fuel rods veuld be at a lov te=perature, consequently, gaseous release vould be expected to censist of ncble gases as halogens vculd be subject to " scrubbing-cut" in the water. The released noble gas is estimated to consist of 10% of the total noble gas content in the da aged rods as an upper value. Thus the esti=ated release vould be 2.2% of the total noble gas content in the core. It is assumed that the noble gas is released at the time of the accident which is assumed to occur at twelve hourt after shutdown of the reactor. The total noble cas radioactivity, a=cunting to

( 15 SECTION 12 - SAFETY ANALYSIS (Contd) 2. (Contd) approximately 520,000 curies would be released to the containment vessel atmosphere. This accident would produce no attendant pres-sure affects within the containment vessel and therefore no un-controlled leakage to the environs. The existence of high radiation levels would be indicated and alar cd by the containment vessel area monitors and by the stack monitor. Consequently, an operator eculd te able to control ti.a subsequent release to the environs in ac-cordance with the pe missible annual average statt,em".:sion rttes. These consequences are insir..ificant in comparison to the consequences described for the " maximum credible accident." G e 6}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20009A182&oldid=4658274"